IR 05000416/2022003
| ML22306A103 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 11/03/2022 |
| From: | Jeffrey Josey NRC/RGN-IV/DORS/PBC |
| To: | Kapellas B Entergy Operations |
| Schaup W | |
| References | |
| IR 2022003 | |
| Download: ML22306A103 (19) | |
Text
November 3, 2022
SUBJECT:
GRAND GULF NUCLEAR STATION - INTEGRATED INSPECTION REPORT 05000416/2022003
Dear Brad Kapellas:
On September 30, 2022, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Grand Gulf Nuclear Station. On October 6, 2022, the NRC inspectors discussed the results of this inspection with Russell Williams, Acting General Manager Plant Operations, and other members of your staff. The results of this inspection are documented in the enclosed report.
Two findings of very low safety significance (Green) are documented in this report. One of these findings involved a violation of NRC requirements. We are treating this violation as a non-cited violation (NCV) consistent with section 2.3.2 of the Enforcement Policy.
A licensee-identified violation which was determined to be Severity Level IV is documented in this report. We are treating this violation as a non-cited violation (NCV) consistent with section 2.3.2 of the Enforcement Policy.
If you contest the violations or the significance or severity of the violations documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; the Director, Office of Enforcement; and the NRC Resident Inspector at Grand Gulf Nuclear Station.
If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; and the NRC Resident Inspector at Grand Gulf Nuclear Station. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Jeffrey E. Josey, Chief Projects Branch C Division of Operating Reactor Safety Docket No. 05000416 License No. NPF-29
Enclosure:
As stated
Inspection Report
Docket Number:
05000416
License Number:
Report Number:
Enterprise Identifier:
I-2022-003-0012
Licensee:
Entergy Operations, Inc.
Facility:
Grand Gulf Nuclear Station
Location:
Port Gibson, MS
Inspection Dates:
July 1 to September 30, 2022
Inspectors:
A. Smallwood, Resident Inspector
T. Steadham, Senior Resident Inspector
Approved By:
Jeffrey E. Josey, Chief
Projects Branch C
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Grand Gulf Nuclear Station, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information. A licensee-identified non-cited violation is documented in report section: 7115
List of Findings and Violations
Failure to Prevent Spurious RCIC Isolation Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000416/2022003-01 Open/Closed None 71111.15 A self-revealed, Green finding and associated non-cited violation of 10 CFR Part 50, appendix B, criterion III, Design Control, was identified when the reactor core isolation cooling system spuriously isolated on a recoverable pressure perturbation in main steam line A. The spurious actuation occurred because the licensee failed to verify or check the adequacy of design for the trip logic signal for the reactor core isolation cooling outboard steam isolation valve.
Failure to Properly Align High Pressure Control Valves Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green FIN 05000416/2022003-02 Open/Closed
[H.4] -
Teamwork 71111.19 A self-revealed, Green finding was identified when the licensee failed to ensure that a design change was developed in accordance with licensee procedure EN-DC-115. Specifically, the licensee failed to perform a thorough review of all the information contained in the document to ensure that it was technically adequate and failed to verify the adequacy of the design. As a result, two high pressure turbine control valve stems sheared causing each valve to fail closed. With two high pressure control valves closed, operators were required to shutdown the plant.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000416/2021-002-00 Core Monitoring System Software Modeling Error Resulted in Conditions Prohibited by Technical Specifications 71153 Closed
PLANT STATUS
Grand Gulf Nuclear Station, Unit 1, began the inspection period shutdown for forced outage (FO) 24-01. On July 3, 2022, operators began a plant startup following completion of maintenance to address the cause of the FO and achieved 77 percent rated thermal power (RTP) on July 4, 2022. On July 4, 2022, operators reduced power to 59 percent RTP due to the failure of one main turbine high pressure control valve. On July 12, 2022, operators shutdown the unit when a second main turbine high pressure control valve failed. On August 5, 2022, operators began a plant startup after completing repairs to the main turbine high pressure control valves. On August 12, 2022, the unit reached 92 percent RTP.
On August 16, 2022, operators reduced power to 57 percent RTP for a rod pattern adjustment.
On August 17, 2022, the unit reached full RTP where it remained at or near until August 19, 2022, when operators manually shutdown the plant because of a failed component in the automatic depressurization system. After repairs were complete operators started up the plant on August 21, 2022. The plant reached 93 percent RTP on August 22, 2022.
On August 23, 2022, operators reduced power to 59 percent RTP for a rod pattern adjustment.
On August 24, 2022, the unit reached 100 percent RTP. On September 17, 2022, operators reduced power to 60 percent RTP for a rod sequence exchange. The unit was returned to 100 percent RTP on September 19, 2022, where it remained at or near for the remainder of the inspection period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed on-site portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.01 - Adverse Weather Protection
Seasonal Extreme Weather Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated readiness for seasonal extreme weather conditions prior to the onset of seasonal hot temperatures for the following systems:
standby service water
standby diesel generators
plant service water
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (2 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
(1)division 1 standby diesel generator air start system on September 9, 2022 (2)division 3 service water while reactor core isolation cooling is out of service for planned maintenance on September 23, 2022
Complete Walkdown Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated system configurations during a complete walkdown of the high-pressure core spray system while reactor core isolation cooling was inoperable on August 10, 2022.
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (5 Samples)
The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:
(1)standby diesel generator building breezeway on August 10, 2022 (2)fire zone yard walkdown on August 19, 2022 (3)division 1 standby diesel generator room on September 9, 2022 (4)division 3 standby diesel generator room on September 16, 2022 (5)standby service water A pump house on September 23, 2022
Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the onsite fire brigade performance during a fire brigade response to report of smoke in the security shift briefing room on September 12, 2022.
71111.07A - Heat Exchanger/Sink Performance
Annual Review (IP Section 03.01) (1 Sample)
The inspectors evaluated readiness and performance of:
- (1) high-pressure core spray pump room cooler on September 12, 2022
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (2 Samples)
- (1) The inspectors observed and evaluated licensed operator performance in the control room during plant shutdown for main turbine high pressure control valve failure on July 12, 2022.
- (2) The inspectors observed and evaluated licensed operator performance in the control room during plant startup from forced outage 24-02 on August 5, 2022.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (3 Samples)
- (1) The inspectors observed and evaluated a licensed operator training evolution in the simulator on August 9, 2022.
- (2) The inspectors observed and evaluated a licensed operator training evolution in the simulator on August 15, 2022.
- (3) The inspectors observed and evaluated a licensed operator requalification scenario in the simulator on September 20, 2022.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (2 Samples)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:
(1)condition report CR-GGN-2022-04789, circulating water pump B trip on August 9, 2022 (2)condition report CR-GGN-2022-07173, service life controls on replacement O-rings for containment airlock air valves on August 9, 2022
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (5 Samples)
The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:
(1)high-pressure core spray while reactor core isolation cooling was isolated on August 9, 2022 (2)protected system verification with radial well 3 out of service on August 19, 2022 (3)protected system lineup with reactor core isolation cooling tagged out for maintenance on September 14, 2022 (4)maintenance risk assessment while radial well 4 out of service for lateral redevelopment on September 23, 2022 (5)protected system lineup while high pressure core spray standby diesel generator unavailable due to maintenance and testing on September 29, 2022
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (5 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
(1)condition report CR-GGN-2022-07617, raised water rod on fuel assembly GEX290 on August 8, 2022 (2)condition report CR-GGN-2022-08821, standby diesel generator starting air leak on September 9, 2022 (3)condition report CR-GGN-2022-08328, automatic depressurization system valve operability with failed relief valve on September 23, 2022 (4)condition report CR-GGN-2022-06950, isolation of reactor core isolation cooling steam line on September 23, 2022 (5)condition report CR-GGN-2022-09366, standby service water head tank level cycling on September 29, 2022
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (1 Sample)
The inspectors evaluated the following temporary or permanent modifications:
(1)engineering change EC-93394, reactor core isolation cooling 5-second time delay on September 27, 2022
71111.19 - Post-Maintenance Testing
Post-Maintenance Test Sample (IP Section 03.01) (8 Samples)
The inspectors evaluated the following post-maintenance testing activities to verify system operability and/or functionality:
(1)work order 52839101, replace/repair containment airlock outer door equalizing valve on August 3, 2022 (2)work order 584914, replace 1B21F125H valve with pre-certified valve on August 24, 2022 (3)work order 531317, replace Namco limit switches on D outboard main steam isolation valve 1B21F028D on August 25, 2022 (4)work order 580099, repair standby diesel generator 11 temperature control valve on August 25, 2022 (5)work order 582291, replace stem on main turbine high pressure control valve A on September 9, 2022 (6)work order 52922542, replace stem nut on 1E51F022D, condensate storage tank suction valve for reactor core isolation cooling pump on September 23, 2022 (7)work order 538760, replace oil in standby service water pump C motor on September 23, 2022 (8)work order 559063, replace 1N11F026A, high-pressure control valve, on September 27, 2022
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (3 Samples)
- (1) The inspectors evaluated forced outage 24-01 activities from July 1 through July 3, 2022. The inspectors completed inspection procedure section 03.01.d. on July 3, 2022.
- (2) The inspectors evaluated forced outage 24-02 activities from July 11 through August 8, 2022. The inspectors completed inspection procedure sections 03.01.b, 03.01.c, and 03.01.d. on August 9, 2022.
- (3) The inspectors evaluated forced outage 24-03 activities from August 19 through August 21, 2022. The inspectors completed inspection procedure sections 03.01.b and 03.01.d. on August 22, 2022.
71111.22 - Surveillance Testing
The inspectors evaluated the following surveillance testing activities to verify system operability and/or functionality:
Surveillance Tests (other) (IP Section 03.01) (4 Samples)
(1)work order 560969, replace valve with pre-certified valve on August 21, 2022 (2)perform procedure 06-OP-1P75-V-0013, standby diesel generator 12 operability verification for limiting condition for operation 1-TS-22-1277 on August 25, 2022 (3)work order 53007300, turbine stop valve trip pressure function test reactor protection system channel B on September 26, 2022 (4)work order 52908522, reactor core isolation cooling overspeed trip test on September 26, 2022
71114.06 - Drill Evaluation
Select Emergency Preparedness Drills and/or Training for Observation (IP Section 03.01) (1 Sample)
- (1) Red team drill on September 21, 2022
Drill/Training Evolution Observation (IP Section 03.02) (1 Sample)
The inspectors evaluated:
(1)classification and notification evaluation during licensed operator requalification simulator training on August 15,
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
MS06: Emergency AC Power Systems (IP Section 02.05)===
- (1) July 1, 2021, through June 30, 2022
MS07: High Pressure Injection Systems (IP Section 02.06) (1 Sample)
- (1) July 1, 2021, through June 30, 2022
MS08: Heat Removal Systems (IP Section 02.07) (1 Sample)
- (1) July 1, 2021, through June 30, 2022
71152A - Annual Follow-up Problem Identification and Resolution Annual Follow-up of Selected Issues (Section 03.03)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
(1)condition report CR-GGN-2007-0378, corrective actions associated with division 1 emergency diesel generator temperature control valve failure on August 19, 2022
71153 - Follow Up of Events and Notices of Enforcement Discretion Event Followup (IP Section 03.01)
- (1) The inspectors evaluated operator response to high pressure turbine control valve A closure on July 6, 2022.
- (2) The inspectors evaluated the licensees response to a failed pressure relief valve and technical specification required shutdown on August 21, 2022.
Event Report (IP Section 03.02) (1 Sample)
The inspectors evaluated the following licensee event reports (LERs):
- (1) LER 05000416/2021-002-00, Core Monitoring System Software Modeling Error Resulted in Conditions Prohibited by Technical Specifications (ML21228A259). The circumstances surrounding this LER and a licensee-identified Severity Level IV non-cited violation is documented in the Inspection Results section of this report.
Personnel Performance (IP Section 03.03) (1 Sample)
- (1) The inspectors evaluated operator response to high pressure turbine control valve D closure on July 11,
INSPECTION RESULTS
Failure to Prevent Spurious RCIC Isolation Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000416/2022003-01 Open/Closed None 71111.15 A self-revealed, Green finding and associated non-cited violation of 10 CFR Part 50, appendix B, criterion III, Design Control, was identified when the reactor core isolation cooling system spuriously isolated on a recoverable pressure perturbation in main steam line A. The spurious actuation occurred because the licensee failed to verify or check the adequacy of design for the trip logic signal for the reactor core isolation cooling outboard steam isolation valve.
Description:
On July 4, 2022, the control room received alarms indicating the reactor core isolation cooling (RCIC) system isolated when the RCIC outboard steam supply isolation valve 1E51F064 automatically closed. RCIC was declared inoperable while the system was isolated. Further investigation found that main steam line A high pressure turbine control valve stem sheared causing a sudden closure of the valve resulting in a pressure perturbation in main steam line A. Since supply steam for RCIC is provided from main steam line A, the pressure perturbation was detected by the RCIC steam line pressure monitoring system which caused 1E51F064 to close and isolate the RCIC system. The RCIC isolation event was documented in the corrective action program as condition report CR-GGN-2022-06950 and the control valve failure was documented in the corrective action program as condition report CR-GGN-2022-06956.
The RCIC steam supply line has two isolation valves, the 1E51F063 RCIC inboard isolation valve and 1E51F064. Transmitters 1E31N083A and 131N083B initiate an isolation trip logic signal for 1E51F063 and 1E51F064, respectively. 1E31N083A had a 5 second time delay but 1E31N083B did not. Due to the pressure perturbation lasting for less than 5 seconds, 1E51F063 did not close; however, 1E51F064 did close. A closure of either valve was sufficient to isolate steam to the RCIC turbine. The inspectors determined that the lack of a time delay for 1E31N083B has existed since plant construction.
As described in Updated Final Safety Analysis Report (UFSAR), section 7.6.1.4.3.3.4.1, the RCIC steam line differential pressure transmitters measure the differential pressure created as steam flows past an elbow in the line, so that the steam flow rate through it can be monitored and used to indicate the presence of a leak (or break). In the presence of a leak, the RCIC system responds by generating the auto-isolation signal. Where required, time delays have been incorporated in this isolation logic to prevent spurious isolation due to pressure spikes which accompany system startups. The original purpose of the time delay was to prevent spurious RCIC isolations during system initiation.
Based on the above information, the inspectors determined that the function of 1E31N083B was to isolate RCIC on a valid leak or break in the RCIC steam supply line. The design of the installed transmitter and the associated trip logic signal was susceptible to spurious actuations as demonstrated by the pressure perturbation when the high-pressure turbine control valve suddenly closed. The RCIC steam supply outboard isolation valve should have remained open to maintain the system operable since no valid leak or line break existed.
The licensee determined that an appropriate solution was to add a 5-second time delay on 1E31N083B to prevent spurious isolations. During a subsequent forced outage, the licensee incorporated this 5-second delay for 1E31N083B using engineering changes EC93394, EC93431, and EC93432, and work orders 58210, and 58211.
Corrective Actions: The licensee added a 5-second time delay to the 1E31N084B transmitter.
Corrective Action References: condition reports CR-GGN-2022-06950 and CR-GGN-2022-06956
Performance Assessment:
The failure to either verify or check the adequacy of design for the trip logic signal for the RCIC outboard steam isolation valve 1E51F064 was contrary to 10 CFR Part 50, appendix B, criterion III and was a performance deficiency. Specifically, the licensee failed to ensure that the parts selected to provide the isolation trip logic signal for 1E51F064 would function to prevent spurious isolations.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to implement this time delay resulted in the isolation of the RCIC system and prevented the automatic initiation of RCIC while the system was isolated.
Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, exhibit 2
- Mitigating Systems Screening Questions. Because the inspectors answered Yes to question number two, a detailed risk evaluation was performed by the senior reactor analyst.
The analyst made the following assumptions related to the subject performance deficiency:
1. The site-specific SPAR model combined with hand calculations were the best
available tools to analyze the risk of this finding.
2. The only initiator that would potentially result from a failure of a single control or stop
valve would be a plant transient.
3. Given the performance deficiency, only the failure of a valve in steam line A would
result in an inadvertent isolation of RCIC.
Using the SPAR model, the analyst determined that the dominant sequence for a transient with loss of RCIC was Sequence 75 (representing 53 percent of the core damage frequency),which resulted from the following:
A general plant transient, modeled with an initiating event frequency of 0.74/year. The analyst noted that this was bounding because very few of the plant transients are initiated by a failure of a turbine control or stop valve.
Failure to maintain condenser vacuum with a probability of 8.70E-04.
Failure of high-pressure core spray with a probability of 2.30E-02.
Failure of RCIC (assumed).
Failure of operators to depressurize the reactor with a probability of 5.02E-04.
Failure of the control rod drive hydraulic system (assumed to fail in the model).
The product of these probabilities combined with the frequency of a transient, resulted in a conditional core damage frequency of 7.43E-09. The analyst adjusted this value for the additional 47 percent of the cutsets and took one-fourth of the risk based on only one of four steam lines would result in the isolation of RCIC. Given that this condition lasted for more than a year, the analyst determined that the incremental conditional core damage probability resulting from this performance deficiency was less than 3.51E-09. Therefore, this finding is of very low safety significance (Green).
Cross-Cutting Aspect: None. This violation is not indicative of current licensee performance; therefore, no cross-cutting aspect is identified.
Enforcement:
Violation: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems and components.
Contrary to the above, until July 30, 2022, the licensee failed to review for suitability of application of parts essential to the safety-related functions of the reactor core isolation cooling system. Specifically, the licensee failed to ensure that the parts selected to provide the isolation trip logic signal for 1E51F064, the reactor core isolation cooling outboard steam isolation valve, would function to prevent spurious isolations.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with section 2.3.2 of the Enforcement Policy.
Failure to Properly Align High Pressure Control Valves Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green FIN 05000416/2022003-02 Open/Closed
[H.4] -
Teamwork 71111.19 A self-revealed, Green finding was identified when the licensee failed to ensure that a design change was developed in accordance with licensee procedure EN-DC-115. Specifically, the licensee failed to perform a thorough review of all the information contained in the document to ensure that it was technically adequate and failed to verify the adequacy of the design. As a result, two high pressure turbine control valve stems sheared causing each valve to fail closed. With two high pressure control valves closed, operators were required to shutdown the plant.
Description:
On July 4, 2022, the control room received alarms indicating a reduction in the steam flow in the A main steam line. Further investigation found that A main steam line high pressure turbine control valve (HPCV) 1N11F026A stem had sheared causing a sudden closure of the valve and a resulting pressure perturbation in A main steam line. The control valve failure was documented in the corrective action program as condition report CR-GGN-2022-06956.
On July 12, 2022, operators observed similar indications for the D main steam line and further investigation found that D main steam line HPCV 1N11F026D failed in the same manner as 1N11F026A. With two main turbine steam lines isolated, plant procedures required operators to perform an orderly plant shutdown that was completed later that night. The D control valve failure was documented in the corrective action program as condition report CR-GGN-2022--07176.
The actuators for all four HPCVs were replaced during refueling outage 23 in the spring 2022.
This work was evaluated under Engineering Change (EC) EC-89459, "TCS HPCV Actuator Replacement/Redesign by Voith," In this EC, the licensee evaluated the effects on the maximum lateral loads of the actuator on the valve stem. Performing this evaluation required the licensee to establish coupling face alignment criteria of the valve and actuator stems to ensure the lateral loads remained within design limits to prevent damage to the stem. For this evaluation, the licensee used the actuator stem to valve stem alignment criteria given by the actuator manufacturer but did not consult the valve manufacturer to obtain their alignment acceptance criteria, which was available to the licensee at the time.
The licensee performed a failure evaluation of both failed valves and determined the cause of failure for both stems was a direct result of an inadequate alignment of the coupling halves.
The licensee determined that the as-found alignment of both 1N11F026A and D exceeded the valve manufacturers acceptance criteria.
Licensee Procedure EN-DC-115, "Engineering Change Process," revision 32, contained two relevant steps that the design reviewer was responsible for:
step 4.2.1: Performing a thorough review of all information contained in the Engineering Change to ensure that the document is technically adequate, procedurally compliant, accurate, and of a quality to warrant approval and issuance
step 4.2.5: Verifying the adequacy of the design by reviewing, confirming or substantiating that the design meets the specified design inputs The inspectors concluded that during the development of EC-89459, the licensee failed to meet both steps when the appropriate coupling face parallelism alignment criteria was not identified.
Corrective Actions: The licensee inspected all four HPCV actuators and valves, replaced all damaged components, redesigned the couplings to reduce stresses on the rod and stem, ensured the couplings were all properly aligned between the actuators and the valve stems, updated vendor documents to include alignment criteria, and revised procedures to ensure that interfacing vendors were included as stakeholders on design changes.
Corrective Action References: condition reports CR-GGN-2022-06956 and CR-GGN-2022-07176
Performance Assessment:
Performance Deficiency: The failure to ensure that EC-89459 was developed in accordance with procedure EN-DC-115 was a performance deficiency. Specifically, the licensee failed to both perform a thorough review of all the information contained in the document to ensure that it was technically adequate and to verify the adequacy of the design Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to ensure that EC89459 was adequately developed contributed to the failure to properly align all four HPCVs. This alignment failure caused two of the HPCVs to fail closed which required a plant shutdown.
Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power.
Because the inspectors answered No to all questions, this performance deficiency screened as Green.
Cross-Cutting Aspect: H.4 - Teamwork: Individuals and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, the failure to maintain strong cross-organizational communications was a direct contributor to provide adequate alignment criteria in preparing the work orders that installed the actuators.
Enforcement:
The inspectors did not identify a violation of regulatory requirements associated with this finding.
Licensee-Identified Non-Cited Violation 71153 This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with section 2.3.2 of the Enforcement Policy.
Violation: The licensee identified a Severity Level IV non-cited violation (NCV) of Technical Specification (TS) 3.2.2, Minimum Critical Power Ratio (MCPR), upon review of operational history and General Electric - Hitachi (GEH) Safety Communication SC 21-04, Fuel Support Side Entry Orifice Meta-Stable for 2 Beam Locations in the BWR/6 Reactors, revision 1.
Technical specification limiting condition for operation (LCO) 3.2.2 requires that all maximum critical power ratios (MCPRs) shall be greater than or equal to the MCPR operating limits specified in the COLR. LCO 3.2.2, condition A, states that for any MCPR not within limits, MCPRs are required to be restored within limits in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. LCO 3.2.2, condition B, states if the required action and associated completion time for condition A is not met, the licensee is required to reduce thermal power to < 21.8 percent rated thermal power (RTP) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Contrary to the above, on multiple occasions between August 13, 2018, and November 6, 2020, MCPR was not within limits for greater than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the licensee did not reduce thermal power to < 21.8 percent RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the failure to meet the required action and associated completion time for condition A.
This violation is being treated as a non-cited violation, consistent with section 2.3.2 of the Enforcement Policy. No additional violations were identified as a result of this LER review.
This LER is closed.
Significance/Severity: No Performance Deficiency. Severity Level IV. The NRC determined this violation was not reasonably foreseeable and preventable by the licensee, and therefore, is not a performance deficiency. Enforcement Policy, section 2.2.4, states that violations with no associated performance deficiency will be dispositioned using traditional enforcement.
Therefore, operating reactor violations with no associated performance deficiencies should be assigned a severity level. The inspectors determined the severity of the violation using section 6 of the NRC Enforcement Policy and determined this issue was Severity Level IV because it most represented the examples in section 6.1.d. The failure to meet the TS LCO and action statement was unknown to the licensee until Safety Communication SC 21-04 was received from GEH, and a review of operational history was performed.
Corrective Action References: condition reports CR-GGN-2021-02742, CR-GGN-2021-03024, CR-GGN-2021-04909, and CR-HQN-2021-01048
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
On October 6, 2022, the inspectors presented the integrated inspection results to Russell Williams, Acting General Manager Plant Operations, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
CR-GGN-
20-00915, 2020-05697, 2020-11531, 2021-08158, 2022-
01110, 2022-06449, 2022-06869, 2022-07513
DWG M1061A
P&I Diagram Standby Service Water System Unit 1
DWG M1061B
P&I Diagram Standby Service Water System USFAR Figure
09.2-001
DWG M1086
P&I Diagram High Pressure Core Spray System Unit 1
DWG M1093A
P&I Diagram HPCS Diesel Generator System
DWG M1093B
P&I Diagram HPCS Diesel Generator System
Drawings
DWG M1093C
P&I Diagram HPCS Diesel Generator System Unit 1
SOI 04-1-01-E22-
High Pressure Core Spray System
130
SOI 04-1-01-P41-
Standby Service Water System
155
Miscellaneous
SOI 04-1-01-P81-
High Pressure Core Spray System Diesel Generator
Work Orders
WO 538760, 583733
Corrective Action
Documents
CR-GGN-
20-12390, 2021-02462, 2021-04968, 2021-07384, 2022-
05887
Work Orders
Miscellaneous
Training Material
No. GSMS-LOR-
370
Loss of Instrument Air/LOCA/Loss of All Level Indication
Procedures
EPP 01-02
Emergency Action Levels
Corrective Action
Documents
CR-GGN-
22-04789, 2022-07173
10538284
Purchase Order
10574643
Purchase Order
55636
Receipt Report
Miscellaneous
58664
Receipt Report
Work Orders
WO 578361, 52920898, 52940066
Corrective Action
Documents
CR-GGN-
22-06950, 2022-07617, 2022-07670, 2022-07719, 2022-
07871, 2022-08328, 2022-08821
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Engineering
Report ECH-NE-
2-00019
Lifted Water Rod on Fuel Assembly
Miscellaneous
GNF Report
007N2461
Grand Gulf Cycle 24 Raised Water Rod Evaluation
EC93342
Engineering
Changes
Work Orders
WO 210, 58211
Corrective Action
Documents
CR-GGN-
20-00915, 2022-05559, 2022-06950, 2022-06956, 2022-
07176, 2022-07818, 2022-09366
Engineering
Changes
TCS HPCV Actuator Replacement/Redesign by Voith
2/23/2022
10537921
Purchase Order
Miscellaneous
55636
Receipt Inspection
Procedures
Engineering Change Process
Work Orders
WO 531314, 531315, 531316, 531317, 538760, 558976, 559062,
559063, 559065, 560969, 580099, 582291, 584914,
2839101, 52841032, 52922542
Procedures
06-OP-1P75-V-
0013
Standby Diesel Generator 12 Operability Verification
05/17/22
Work Orders
WO 560969, 52935628, 53007300
71151
Miscellaneous
Performance Indicator Technique Data Sheets for Third
Quarter 2021 through Second Quarter 2022
Corrective Action
Documents
CR-GGN-
2007-00378, 2022-05734
04-1-01-P75-1
Standby Diesel Generator System Operating Instruction
117, 118
Procedures
04-1-02-1H22-
P400
Alarm Response instruction Panel No. 1H22-P400
20, 121